核反应堆堆芯吊篮压紧弹簧松弛特性

doi:10.16183/j.cnki.jsjtu.2023.024

上海交通大学学报 ›› 2024, Vol. 58 ›› Issue (8): 1290-1296.doi: 10.16183/j.cnki.jsjtu.2023.024

• 机械与动力工程 • 上一篇

核反应堆堆芯吊篮压紧弹簧松弛特性

杨泰波¹^,², 刘佳鑫², 彭志科³(), 罗能², 刘才学²

1.上海交通大学机械与动力工程学院,上海 200240
2.中国核动力研究设计院,成都 610213
3.宁夏大学前沿交叉学院,银川 750021

收稿日期:2023-01-19 修回日期:2023-04-23 接受日期:2023-04-24 出版日期:2024-08-28 发布日期:2024-08-27
通讯作者: 彭志科,教授;E-mail:z.peng@sjtu.edu.cn.
作者简介:杨泰波(1986-),副研究员,从事反应堆故障诊断与健康管理技术研究.

Relaxation Characteristics of Hold-Down Spring of Core Support Barrel in Nuclear Reactor

YANG Taibo¹^,², LIU Jiaxin², PENG Zhike³(), LUO Neng², LIU Caixue²

1. School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200240, China
2. Nuclear Power Institute of China, Chengdu 610213, China
3. School of Advanced Interdisciplinary Studies, Ningxia University, Yinchuan 750021, China

Received:2023-01-19 Revised:2023-04-23 Accepted:2023-04-24 Online:2024-08-28 Published:2024-08-27

摘要/Abstract

摘要：

核反应堆中堆芯吊篮的稳定性是反应堆正常运行的保障和前提.压紧弹簧作为堆芯吊篮的约束与支撑直接影响堆芯吊篮的振动特性.在高温、辐射等恶劣工况下,压紧弹簧力学性能会发生如应力松弛的退化,影响堆芯吊篮振动以及反应堆运行安全,因此需要对堆芯吊篮压紧弹簧松弛特性展开研究.首先,建立堆芯吊篮及压紧弹簧在内的装配体有限元模型,利用湿模态法分析并获得压紧弹簧松弛情况下的堆芯吊篮梁式模态振动频率的变化.然后,开展堆芯吊篮压紧弹簧松弛试验,试验结果验证了模型的有效性.进一步利用仿真模型生成更多压紧弹簧松弛影响下的堆芯吊篮梁式模态频率数据,结合试验数据建立堆芯吊篮压紧弹簧松弛程度识别的数学模型,为反应堆堆芯吊篮压紧弹簧松弛劣化监测奠定了技术基础.

关键词: 堆芯吊篮, 压紧弹簧, 松弛特性, 模态分析, 松弛试验

Abstract:

The stability of the core support barrel (CSB) in a nuclear reactor is the guarantee and premise for the normal operation of the reactor core. As the constraint and support of the CSB, the hold-down spring (HDS) directly affects the vibration characteristics of the CSB. Under harsh conditions such as high temperature and radiation, degradation such as stress relaxation may occur in terms of the performance of the HDS, which will affect the vibration of the CSB and the operation security of the reactor. Therefore, it is necessary to study the relaxation characteristics of the HDS of the CSB. First, a finite element model including the CSB and the HDS is established. The wet modal method is adapted to analyze and obtain the change of the beam mode frequency of the CSB with the relaxation degradation of the HDS. Then, a relaxation test of the HDS of the CSB is conducted to verify the simulation results. The simulation model is further used to generate more beam mode frequency data of the CSB under the impact of different relaxation degrees of the HDS. Finally, combined with simulation and test data, a mathematical model for identifying the relaxation degree of the HDS of the CSB is established, which lays a technical foundation for monitoring the relaxation and degradation of the HDS of the CSB in nuclear reactor.

Key words: core support barrel (CSB), hold-down spring (HDS), relaxation characteristics, modal analysis, relaxation test

中图分类号:

TL351

杨泰波, 刘佳鑫, 彭志科, 罗能, 刘才学. 核反应堆堆芯吊篮压紧弹簧松弛特性[J]. 上海交通大学学报, 2024, 58(8): 1290-1296.

YANG Taibo, LIU Jiaxin, PENG Zhike, LUO Neng, LIU Caixue. Relaxation Characteristics of Hold-Down Spring of Core Support Barrel in Nuclear Reactor[J]. Journal of Shanghai Jiao Tong University, 2024, 58(8): 1290-1296.

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参考文献 16

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试验顺序	试验内容	试验模拟方法
1	压紧弹簧正常状态下堆芯吊篮振动特性试验	正常刚度弹簧
2	不同刚度压紧弹簧堆芯吊篮振动特性试验	压紧弹簧减小15%刚度
3	不同刚度压紧弹簧堆芯吊篮振动特性试验	压紧弹簧减小30%刚度
4	不同刚度压紧弹簧堆芯吊篮振动特性试验	压紧弹簧减小50%刚度

序号	松弛程度/%	堆芯吊篮一阶梁式频率
序号	松弛程度/%	试验值/Hz	仿真值/Hz	误差/%
1	0	10.4	10.62	2.12
2	15	9.4	9.56	1.70
3	30	8.1	8.23	1.60
4	50	6.1	5.98	-1.97

序号	压紧弹簧松弛程度/%	堆芯吊篮一阶梁式频率
序号	压紧弹簧松弛程度/%	仿真值/Hz	试验值/Hz	拟合值/Hz
1	0	10.62	10.4	10.400
2	5	10.31		10.110
3	10	9.95		9.770
4	15	9.56	9.4	9.400
5	20	9.17		9.020
6	30	8.23	8.1	8.100
7	40	7.16		7.115
8	50	5.98	6.1	6.100
9	60	4.76		5.005
10	70	3.39		3.760

核反应堆堆芯吊篮压紧弹簧松弛特性

Relaxation Characteristics of Hold-Down Spring of Core Support Barrel in Nuclear Reactor

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